Text

Amend 33 to License NPF-86 Revising TS to Permit Operation of Plant Core W/Expanded Axial Flux Difference Band from That Currently Permitted
	ML20078L902
Person / Time
Site:	Seabrook
Issue date:	11/23/1994
From:	Mckee P; Office of Nuclear Reactor Regulation
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gnarco 2*,

UNITED STATES er j

j NUCLEAR REGULATORY COMMISSION 1"

WASHINGTON, D.C. 20555-0001

%...../

NORTH ATLANTIC ENERGY SERVICE CORPORATION. ET AL*

DOCKET NO. 50-443 SEABROOK STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 33 License No. NPF-86 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by North Atlantic Energy Service Corporation, et al. (the licensee), dated November 23, 1993, as supplemented by letter dated August 15, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be

^

conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements i

have been satisfied.

t

North Atlantic Energy Service Company (NAESCO) is authorized to act as a.;ent for the: North Atlantic Energy Corporation, Canal Electric Company, The Connecticut Light and Power Company, Great Bay Power Corporation, Hudson Light and Power Department, Massachusetts Municipal Wholesale Electric Company, Montaup Electric Company, New England Power Company, New Hampshire Electric Cooperative, Inc., Taunton Municipal Light Plant, and The United Illuminating Company, and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

9412010210 941123 PDR ADOCK 05000443 p

PDR

+

' 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2,C.(2) of Facility Operating License No. NPF-86 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 33, and the Environmental Protection Plan contained in Appendix B are incorporated into Facility License No.

NPF-86.

NAESCO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

I 3.

This license amendment is offective as of the date of its issuance, to be implemented before startup from the fourth refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION Phillip F. McKee, Director Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 23, 1994 i

i i

. +,.

iITACHMENT TO LICENSE AMENDMENT NO. 33 FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 I

Replace the following pages of Appendix A, Technical Specifications, with the attached pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Overleaf pages have been provided.

l Remove Insert 1-1*

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3/4 1-20 3/4 1-20 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 1

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l B 3/4 3-4 B3/4 3-4 5-9 5-9 5-10*

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e-'

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17 "

4 41.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.

ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.

AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

SEABROOK - UNIT 1 1-1

DEFINITIONS 1

LONTAINMENT INTEGRITY j

1.7 CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g.,

welds, bellows, or 0-rings) is OPERABIE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel nead removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) provides core operating limits l

for the current operating reload cycle. The cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.8.1.6.

Plant operation within these operating limits is addressed in individual specifications.

DIGITAL CHANNEL OPERATIONAL TEST 1.11 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and/or injecting simulated process data to verify OPERABILITY of alarm and/or trip functions.

The Digital Channel Operational Test definition is only applicable to the Radiation Monitoring Equipment.

SEABROOK - UNIT 1 1-2 Amendment No. 33

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,,) shall not exceed the limits shown in i

Figure 2.1-1 for four-loop operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Specification 6.6.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.6.

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.6.

1 l

i SEABROOK - UNIT 1 2-1

NOTE:

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION NO TEXT WILL PRINT l

SEABROOK - UNIT 1 2-2 Amendment No. 33

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint valutt shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in lable 3.3-1.

ACTION:

a.

With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip l

Setpoint value.

b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value column of Table 2.2-1, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + S s TA Where:

Z-The value from Column Z of Table 2.2-1 for the affected

channel, l

R-The "as measured" value (in percent span) of rack error for the affected channel, S-Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel.

SEABROOK UNIT 1 2-3

TABLE 2.2-l~

h REACTOR TRIP SYSTEM IN3TRUMENTATION TRIP SETPOINTS 8

x SENSOR TOTAL ERROR i

g FUNCTIONAL UNIT ALLOWANCE (TA) 1 (S)

TRIP SETPOINT ALLOWABLE VALUE

[

1.

Manual Reactor Trip N.A.

N.A.

2.

Power Range, Neutron Flux a.

High Setpoint 7.5 4.56 1.42 s109% of RTP*

slll.1% of RTP*

b.

Low Setpoint 8.3 4.56 1.42 s25% of RTP*

s27.1% of RTP*

3.

Power Range, Neutron Flux, 1.6 0.5 0

s5% of RTP* with s6.3% of RTP* with High Positive Rate a time constant a time constant 22 seconds 22 seconds 4.

Power Range, Neutron Flux, 1.6 0.5 0

55% of RTP* with s6.3% of RTP* with High Negative Rate a time constant a time constant 22 seconds 22 seconds 5.

Intermediate Range, 17.0 8.41 0

s25% of RTP*

s31.1% of RTP*

Neutron Flux 5

5 6.

Source Range, Neutron Flux 17.0 10.01 0

s10 cps

$1.6 x 10 cps 7.

Overtemperature AT N.A N.A N.A See Note 1 See Note 2 8.

Overpower AT N.A N.A N.A See Note 3 See Note 4 E

a 9.

Pressurizer Pressure - Low N.A N.A N.A 21945 psig 21,933 psig 10.

Pressurizer Pressure - High N.A N.A N.A s2385 psig

$2,397 psig

=

P

RTP = RATED THERMAL POWER U

M TABLE 2.2-1 (continued) g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS k

SENSOR i

TOTAL ERROR g

FUNCTIONAL UNIT ALLOWANCE fTA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE

[

11.

Pressurizer Water Level - High 8.0 4.20 0.84 592% of instrument s93.75% of instrument span span 12.

Reactor Ccolant Flow - Low 2.5 1.9 0.6 290% of measured 289.3% of measured loop flow loop flow 13.

5 team Generator Water 14.0 12.53 0.55 214.0% of narrow 212.6% of narrow Level Low - Low range instrument range instrument span span 14.

Undervoltage - Reactor 15.0 1.39 0

210,200 volts 29,822 volts Coolant Pumps 15.

Underfrequency - Reactor 2.9 0

0 255.5 Hz 255.3 Hz Coolant Pumps 16.

Turbine Trip I

a.

Low Fluid Oil Pressure N.A.

N.A.

2500 psig 2450 psig b.

Turbine Stop Valve N.A.

N.A.

21% open 21% open Closure 17.

Safety Injection Input N.A.

N.A.

from ESF g

an O

I

M TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m

k SENSOR i

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE fTA)

Z fS)

TRIP SETPOINT ALLOWABLE VALUE 18.

Reactor Trip System

~

Interlocks a.

Intermediate Range N.A.

N.A.

21 x 10

amp 26 x 10'" amp -

Neutron Flux, P-6 b.

Low Power Reactor Trips Block, P-7

1) P-10 input N.A.

N.A.

s10% of RTP*

s12.1% of RTP*

2)

P-13 iniut N.A.

N.A.

s10% RTP* Turbine

$12.3% of RTP* Turbine l

'P Impulse Pressure-Impulse Pressure l

Equivalent Equivalent c.

Power Range lieutron N.A.

N.A.

s50% of RTP*

s52.1% of RTP*

Flux,-P-8 d.

Power Range Neutron N.A.

N.A.

s20% of RTP*

$22.1% of RTP*

Flux, P-9 e.

Power Range Neutron N.A.

N.A.

210% of RTP*

27.9% of RTP*

Flux, P-10 f.

Turbine Impulse Chamber.

N.A.

s10% RTP* Turbine s12.3% RTP* Turbine Pressure, P-13 Impulse Pressure Impulse Pressure Equivalent Equivalent 19.

Reactor Trip Breakers N.A.

N.A.

20. Automatic Trip and Interlock N.A.

N.A.

M.A._N.A.

N.A.

Logic

RTP = RATED THERMAL POWER

,vc.

...m,-

,m

.,,,e.

m

M TABLE 2.2-1 (Continued)

~

TABLE NOTATIONS 8

n NOTE 1: OVERTEMPERATURE AT

\\

s 7 5) s % (K - K fl + r S)'8) [T III

+

- T'] + K (P - P') - f (AI)}

+r (1

i 2

3 g, 7 3) 3 3

Where:

AT Measured AT by RTD Instrumentation; 1+TS Lead-lag compensator on measured AT; 3

I+7S2 Time constants utilized in lead-lag compensator for AT, r 2 8 s.

r,72 i

i 72 s 3 s; j

1 Lag compensator on measured AT; 1+TS 4

3 Time constants utilized in the lag compensator for AT, r3 - O s; T

3 Indicated AT at RATED THERMAL POWER; AT o

l Value specified in the COLR; K,

=

Value specified in the COLR; K

2 The function generated by the lead-lag compensator for T,

I + r,S 1+TS dynamic compensation; 3

k Time constants utilized in lead-lag compensator for T,,,

7 2 33 s, i,r3 t

4 h

73 s 4 s; e

E.

T Average temperature, "F;

z:

P 1

Lag compensator on measured T L

1+7S g

6 Time constant utilized in the measured T,, lag compensator, 76 - O s; 7

6

E TABLE 2.2-1 (Continued)

TABLE NOTATIONS 8

3 NOTE 1:

(Continued) g T'

s 588.5'F (Nominal T,, at RATED THERMAL POWER);

[

K Value specified in COLR; l

3 Pressurizer pressure, psig; P

P' 2235 psig (Nominal RCS operating pressure);

Laplace transform operator, s;

S and f (AI) is a function of the indicated difference between top and bottom detectors of the i

power-range neutron ion chambers'as specified in the COLR.

'?

oo NOTE 2:

Cycle dependent values for the channel's Allowable Value are specified in the COLR.-

$a 2

a

.F 0

M TABLE 2.2-1 (Continued) g TABLE NOTATIONS (Continued) k NOTE 3: OVERPOWER AT g

AT (1 + T 1)

(1)

(7 5)

(1) m (1 + 7 S) (1 + T 5) s ATo (K

-K (1 < 7 S) (1 + 7 S) T - K6 [T (1 + 7 S) - T"] - f (M))

3 7

4 3

2 2

3 7

6 6

Where:

AT As defined in Note 1, 1+rS As defined in Note 1, i

1+TS2 As defined in Note 1, T, r 3

2 1

As defined in Note 1, 1+753 i*

As defined in Note 1, T

3 As defined in Note 1, AT o

Value specified in the COLR, K

4 Value specified in the COLR, K

3 The function generated by the rate-lag compensator for T,, dynamic 7S 7

1+7S compensation, 7

Time constants utilized in rate-lag compensator for T,,, 77;t 10 s, 7

g 7

1 As defined in Note 1, 2

1+7S6 c+

2 r,

As defined in Note 1,

?

O

M TABLE 2.2-1 (Continued) g TABLE NOTATIONS (Continued) l NOTE 3:

(Continued)

Value specified in COLR, l

K.

c 5

T As defined in Note 1,

~

T" Indicated T at RATED THERMAL POWER (Calibration temperature for AT instrumentaUon,s588.5*F),

S As defined in Note 1, and f (AI) - A function of the indicated difference between the top and bottom detectors of the z

power-range neutron ion chambers as specified in the COLR.

NOTE 4:

Cycle dependent values for the channel's Allowable Value are specified in the COLR.

'?

E 5a Ra E

+

A

2.1 SAFETY LIMITS l

I BASES 2.1.1 REACTOR CORE I

t The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation that would result in the release of fission j

products to the reactor coolant. Overheating of the fuel cladding is prevented q

by restricting fuel operation to within the nucleate boiling regime where the i

heat transfer coefficient is large and the cladding surface temperature is l

slightly above the coolant saturation temperature.

i Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and, therefore, THERMAL POWER and reactor coolant temperature and pressure have been related to DNB. This relation has been developed to predict the DNB flux and l

the location of DNB for axially uniform and nonuniform heat flux distributions.

l The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux l

that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.

The DNB design basis is as follows:

uncertainties in the DNBR correlation, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with 95 percent confidence level that DNB will not occur on the most limiting fuel rod during Condition I and II events. This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

In addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit value, or the average l

enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy rise hot channel factor Fl,, at RATED THERMAL POWER, of 1.65.

The value of Fl, at' reduced power is assumed to vary according to the expression:

l 4

Fl, - 1.65 [1+ 0.3 (1-P)]

Where P is the fraction of RATED THERMAL POWER.

i ThisexpressionconservativelyboundsthecyclespecificlimitsonF"Usin specified in Technical Specification 3/4.2.3 and the COLR. The Safety Lim Figure 2.1-1 are also based on a reference cosine axial power shape with a peak of 1.55.

SEABROOK - UNIT 1 B 2-1 Amendment No. 33

SAFETY LIMITS ~

BASES 2.1.1 REACTOR CORE (Continued)

The resulting heat flux conditions are more limiting than those calculated for the range of all control rods fully withdrawn to the maximum allowable l

control rod insertion, assuming the axial power imbalance is within the limits of the f (AI) and f (al). functions of the Overtemperature and Overpower AT 3

trips. When the axbal power imbalance is not within the tolerance,.the axial i

power imbalance effect on the Overtemperature AI and Overpower AT trips will reduce the setpoints to provide protection consistent with core safety limits for cycle specific power distribution.

i 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor ce91 ant from reaching the-containment atmosphere.

The reactor vessel, pressurizer, and the RCS piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants, which-permits a maximum transient pressure of 110% (2735 psig) of design pressure.:

i The Safety Limit of'2735 psig is, therefore, consistent with the design criteria and associated Code requirements.

The entire RCS is hydrotested at 125% (3110 psig) of design pressure to demonstrate integrity prior to initial operation.

P i

t b

i I

?

SEABROOK UNIT 1 B 2-2 Amendment No. 33

LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Intermediate and Source Rance. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup t, mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron F aReactortripatabout10]uxchannels. The Source Range channels will initiate counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemoerature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), pressure is within the range between the Pressurizer High and Low Pressure trips and power is less than the Overpower AT trip setpoint.

The Setpoint is automatically varicJ with: (1) coolant temperature to correct for temperature induced changes

~

in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature A7 trip, and provides a backup to the High Neutron Flux trip.

The Setpoint is autouatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, (2) rate of change of temperature for l

dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution to ensure that the allowable heat l

generation rate (Kw/ft) is not exceeded.

The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

SEABROOK - UNIT 1 B 2-5 Amendment No. 33

9 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip, thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure that could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Erislyrizer Water level The Pressurizer High Water Level trip is provided to prevent water relief l

through the pressurizer safety valves. On decreasing power, the Pressurizer-High Water Level trip is automatically blocked by P-7 (a power level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full-power equivalent); and on increasing power, the Pressurizer High Water Level trip is automatically reinstated by P-7.

Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB l

by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 50% of RATED THERMAL POWER), an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

Conversely, on decreasing power between P-8 and the P-7, an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

l Steam Generator Water level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch I

resulting from loss of normal feedwater. The specified Setpoint provides allowances for starting delays of the Emergency Feedwater System.

SEABROOK - UNIT 1 B 2-6

REACTIVITY CONTROL SYSTEMS MOVABLE CONTROL ASSEMBLIES POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION t

3.1.3.3.One digital rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod t

position within i 12 steps for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3* **, 4* **, and 5* **

.i ACJ1@:

With less than the above required position indicator (s) OPERABLE, immediately

+

open the Reactor Trip System breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full range of rod travel at least once per 18 months.

L f

t

With the Reactor Trip System breakers in the closed position.

- See Special Test Exceptions Specification 3.10.5 SEABROOK - UNIT 1 3/4 1-19

REACTIVITY CONTROL SYSTEMS MOVABLE CONTROL ASSEMBLIES R0D DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod drop time from the mechanical fully withdrawn position shall be less than or equal to 2.4 l

seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a.

T,, for each loop greater than or equal to 551*F, and b.

All reactor coolant pumps operating.

APPLICABILITY: MODES I and 2.

ACTION:

With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE REOUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:

a.

For all rods following each removal of the reactor vessel head, b.

For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System that could affect the drop time of those specific rods, and c.

At least once per 18 months.

i I

r f

l SEABROOK - UNIT 1 3/4 1-20 Amendment No. 33 1

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a.

The limits specified in the COLR, with the Fixed Incore Detector (FIDS) Alarm OPERABLE, or b.

Tha limits specified in the COLR, when the FIDS Alarm is inoperable.

APPLICABILITY:

MODE I above 50% RATED THERMAL POWER.

ACTION:

a.

With the indicated AFD* outside of the applicable limits specified

)

I in the COLR:

1.

Either restore the indicated AFD to within the COLR specified limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -

)

High Setpoints to less than or equal to 55% of RATED THERMAL 2

POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and 3.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limitt specified in the COLR.

)

b.

With an OPERABLE FIDS Alarm exceeding in limit:

l 1.

Comply with the AFD limits specified in the COLR for operation with the FIDS Alarm inoperable within 15 minutes and, 2.

Verify THERMAL POWER is less than the maximum power limit established by Surveillance Requirement 4.2.1.2 within 15 minutes and, 3.

Identify and correct the cause of the FIDS Alarm prior to operation beyond the limits specified in the COLR for operation with the FIDS Alarm inoperable.

I c.

With the FIDS Alarm inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , l

1.

Comply with the AFD limits specified in the COLR for operation

)

with the FIDS Alarm inoperable, and 2.

Verify THERMAL POWER is less than the maximum power limit established by Surveillance Requirement 4.2.1.2.

The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.

SEABROOK - UNIT 1 3/4 2-1 Amendment No. 33

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE SURVEILLANCE RE0VIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE, and b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 At least once per 31 EFPD determine the maximum allowed power for operation with the FIDS Alarm inoperable by comparing F (Z) to the a

F,(Z) limit established for operation with the FIDS Alarm inoperable.

SEABROOK - UNIT 1 3/4 2-2 Amendment No. 33

1 PAGE INTENTIONALLY BLANK 1

i l

1 i

1 l

SEABROOK - UNIT 1 3/4 2-3 Amendment No. 9

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F.JZ)

LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

a F (Z) s Fy" K(Z) for P > 0.5 a

P F (Z) s F"a'" K(Z) for P s 0.5 a

.5 Where:

P=

THERMAL POWER

, and RATED THERMAL POWER FS"

=

the F limit at RATED THERMAL POWER (RTP) speciliedintheCOLR,and l

K(Z) the normalized F (Z) as a function of core height

=

a as specified in the COLR.

APPLICABILITY: MODE 1.

1 ACTION:

a.

With F (Z) exceeding its limit:

a 1.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limitwithin15minutesandsimilarlyreducet$ePowerRange Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1%

F (Z) exceeds the limit, and a

2.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased, provided F (Z) is demonstrated through incore q

mapping to be within its limit.

l SEABROOK - UNIT 1 3/4 2-4 Amendment No. 33

4 t

PAGE INTENTIONALLY BLANK I

i I

1 i

I i

SEABROOK - UNIT 1 3/4 2-5 Amendment No. 9 1

~

POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR - F.JZ)

LIMITING CONDITION FOR OPERATION 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F (Z) shall be demonstrated to be within its limits prior to a

operation above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:

a.

I! sing the Incore Detector System to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

Increasing the measured F Z) component of the power distribution b.

map by 3% to account for,(anufacturing tolerances and further m

increasing the value by 5% when using the movable incore detectors or 5.21% when using the fixed incore detectors, to account for measurement uncertainties.

4.2.2.3 The limits of Specification 3.2.2 are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

1)

Lower core region from 0 to 15%, inclusive.

2)

Upper core region from 85 to 100%, inclusive.

4.2.2.4 Each fixed incore detector alarm setpoint shall be updated at least once per 31 EFPD. The alarm setpoints will be based on the latest available power distribution, so that the alarm setpoint does not exceed the F (Z) limit defined in Technical Specification 3.2.2.

a SEABROOK - UNIT 1 3/4 2-6 Amendment No. 33

PAGE INTENTIONALLY BLANK t

l SEABROOK - UNIT I 3/4 2-7 Amendment No. 33 I

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 FE shall be less than the limits specified in the COLA.

APPLICABILITY: MODE 1.

ACTION:

With FE exceeding its limit:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required b ACTION a.,

above; THERMAL POWER may then be increased, provided F is demonstrated through incore mapping to be within its 1 it.

SURVEILLANCE RE0VIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F5_ shall be demonstrated to be within its limit prior to operation above 75% MTED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:

a.

Using the Incore Detector System to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER.

b.

Using the measured value of FE which does not include an allowance for measurement uncertainty.

SEABROOK - UNIT 1 3/4 2-8 Amendment No. 33

POWER DISTRIBUTION LIMITS 3/4.2.4 OVADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION j

i 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *.

AGHQff:

With the QUADRANT POWER TILT RATIO determined to exceed 1.02:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of I and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

e b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 7 days thereafter, verify that F,irements (Z) and FlY.are within their limits by performing Surveillance Requ 4.

2.2 and 4.2.3.2.

THERMAL POWER and setpoint reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3.

SURVEILLANCE RE0VIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a.

Calculating the ratio at least once per 7 days when the alarn. is OPERABLE, and b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the Incore Detector System to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by either:

a.

Using the four pairs of symmetric detector locations or b.

Using the Incore Detector System to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.

See Special Test Exceptions Specification 3.10.2 SEABROOK - UNIT 1 3/4 2-9 Amendment No. 33 1

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS l

i LIMITING CONDITION FOR OPERATION 7

t 3.2.5 The following DNB-related parameters shall be maintained within the following limits:

Reactor Coolant System T,,, s 594.3*F a.

b.

Pressurizer Pressure, 2 2185 psig*

c.

Reactor Coolant System Flow shall be:

1.

2 382,800 gpm**; and, 2.

2 392,800 gpm***

APPLICABILITY:

MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits it least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The RCS flow rate indicators shall be subjected to CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by a precision heat balance measurement to be within its limit prior to operation above 95% of RATED THERMAL POWER after each fuel loading. The provisions of Specification 4.0.4 are not applicable for entry into MODE 1.

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10%

of RATED THERMAL POWER.

- Thermal Design Flow. An allowance for measurement uncertainty shall be made when comparing measured flow to Thermal Design Flow.

- - Minimum measured flow used in the Revised Thermal Design Procedure.

SEABROOK - UNIT 1 3/4 2-10 Amendment No. 33 l

i

TABLE 4.3-1 (Continued)

BEACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

9 g;

TRIP g3 ANALOG ACTUATING MODES FOR rs CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERAT'.ONAL ACTUATION SURVEILLANCE gi FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TESTIS REQUIRED

(( Reactor Trip System Interlocks (Continued) e.

Power Range Neutron Flux, P-10 N.A.

R(4)

R N.A.

N.A.

1, 2 f.

Turbine Impulse Chamber Pressure, P-13 N.A.

R R

N.A.

I

19. Reactor Trip Breaker N.A.

N.A.

M(7,11)

N.A.

1, 2, 3*,

4*, 5*

20. Automatic Trip and Interlock N.A.

N.A.

M(7) 1, 2, 3*,

y, Logic 4*,

5*

~~ 21. Reactor Trip Bypass Breaker N.A.

N.A.

M(7, 14),

N.A.

1, 2, 3 *,

R(15) 4*,

5*

I e

n

-,.,n-

TABLE 4.3-1 (Continued)

TABLE NOTATIONS

0nly if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.

- Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

- - Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1)

If not performed in previous 31 days.

(2)

Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3)

Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 50% of RATED THERMAL POWER.

Recalibrate if the absolute difference is l

greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

For the purposes of this surveillance requirement, monthly shall mean at least once per 31 EFPD.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

Initial plateau curves shall be measured for each detector.

Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

For the purposes of this surveillance requirement, quarterly shall mean at least once per 92 EFPD.

(7)

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8)

(Not used)

(9)

Surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

SEABROOK - UNIT 1 3/4 3-12 Amendment No. 33

TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued) l a.

The inoperable channel is placed in the tripped condition within I hour, and b.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing of other chancils per Specification 4.3.2.1.

ACTION 19 - With less than the Minimum Nur*.r of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation ei the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

f ACTION 20 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 21 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 22 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.

ACTION 23 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated valve inoper-able and take the ACTION required by Specification 3.7.1.5.

l l

l I

I SEABROOK - UNIT 1 3/4 3-23

l y

TABLE 3.3-4 t

en ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i

SENSOR c:

TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE

~

1.

Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generator, Phase "A" Isolation, Containment Ventilation Isolation, and Emergency Feedwater, Service Water to Secondary Component ~

Cooling Water Isolation, CBA Emergency Fan / Filter Actuation, and Latching Relay).

a.

Manual Initiation N.A.

N.A.

l b.

Automatic Actuation Logic N.A.

N.A.

c.

Containment Pressure--Hi-1 4.2 0.71 1.67 s 4.3 psig s 5.3 psig d.

Pressurizer Pressure--Low N.A.

N.A.

2 1800 psig 2 1786 psig l

e.

Steam Line Pressure--Low 13.1 10.71 1.63 2 585 psig 2 568 psig*

2.

Containment Spray I

a.

Manual Initiation N.A.

N.A.

E a

b.

Automatic Actuation Logic N.A.

N.A.

E and Actuation Relays R5 c.

Containment Pressure--Hi-3 3.0 0.71 1.67 s 18.0 psig s 18.7 psig E

g TABLE 3.3-4 (Continued)

E ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS 8^

SENSOR TOTAL-ERROR E

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETP0 INT ALLOWABLE VALUE 7.

Emergency Feedwater a.

Manual Initiation (1) Motor driven pump N.A.

N.A.

(2)

Turbine driven pump N.A.

N.A.

l b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays c.

Steam Generator Water 14.0 12.53 0.55 2 14.0% of 212.6% of narrow y

Level--Low-Low narrow range range instrument y

Start Motor-Driven Pump instrument span.

and Start Turbine-Driven span.

Pump i

d.

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Start Motor-Driven Pump Allowable Values.

l and Turbine-Driven Pump e.

Loss-of-Offsite Power See Item 9. for loss-of-Offsite Power Setpoints and Allowable Values.

Start Motor-Driven Pump and Turbine-Driven Pump l

i 8.

Automatic Switchover to Containment Sump a.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays

{

b.

RWST Level--Low-Low 2.75 1.0 1.8 2122,525 gals.

2121,609 gals.

Coincident With Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

_ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _._-_________

v, TABLE 3.3-4 (Continued) 9 E

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

^

SENSOR TOTAL ERROR E

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE Z

9.

Loss of Power (Start Emergency Feedwater) a.

4.16 kV Bus E5 and E6 N.A.

N.A.

2 2975 2 2908 volts Loss of Voltage volts with with a s 1.315 a s 1.20 second time second time delay.

delay.

b.

4.16 kV Bus E5 and E6 N.A.

N.A.

2 3933 volts 2 3902 volts g

Degraded Voltage with a s 10 with a s 10.96 second time second time delay.

delay.

Coincident with:

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

10. Engineered Safety Features Actuation System Interlocks a.

Pressurizer Pressure, P-11 N.A.

N.A.

s 1950 psig s 1962 psig l

b.

Reactor Trip, P-4 N.A.

N.A.

=

c.

Steam Generator Water Level, See Item 5. above for all Steam Generator Water Level Trip g

P-14 Setpoints and Allowable Values.

E I

l l

TABLE 4.3-3 y,

t m

I U

RADIATION MONITORING INSTRUMENTATON FOR PLANT fy OPERATIONS SURVEILLANCE RE00 REMENTS i

DIGITAL CHANNEL MODES FOR WHICH cz CHANNEL CHANNEL OPERATIONAL SURVEILLANCE Z

FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED l

1. Containment j

a. Containment - Post LOCA -

l Area Monitor S

R Q

All

b. RCS Leakage Detection

1) Particulate Radio-S R

Q 1, 2, 3, 4 activity

2) Gaseous Radioactivity S

R Q

1, 2, 3, 4 l

2. Containment Ventilation Isolation l

a. On Line Purge Monitor S

R Q

1, 2, 3, 4 w

l 1

b. Manipulator Crane Area S

R Q

6#

Monitor j

w fE

3. Main Steam Line S

R Q

1, 2, 3, 4 1

4. Fuel Storage Pool Areas l

a. Radioactivity-High-l Gaseous Radioactivity S

R Q

I j

5. Control Roam Isolation

a. Air Intake Radiation Level

1) East Air Intake S

R Q

All

2) West Air Intake S

R Q

All g

l[

6. Primary Component Cooling Water s

a. Loop A S

R Q

All g

b. Loop B S

R Q

All g

TABLE NOTATIONS

With irradiated fuel in the fuel storage pool areas.

g

During CORE ALTERNATIONS or movement of irtadiated fuel within the containment.

INSTRUMENTATION MONITORING INSTRUMENTATION INCORE DETECTOR SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.2 The Incore Detector System shall be OPERABLE with:

a. At least 75% of the detector locations and,

b. A minimum of two detector locations per core quadrant,

c. An OPERABLE incore detector location consist of a fuel assembly containing a fixed detector string with a minimum of three OPERABLE detectors or an OPERABLE movable incore detector capable of mapping the location.

APPLICABILITY:

When the Incore Detector System is used for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO.-or

c. Measurement of FL and F,(Z), or

d. Input into the FIDS Alarm ACILQH:

With the Incore Detector System inoperableapplicablemonitoringorcalibrationfunctIons.n do The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE0VIREMENTS (Plant procedures are used to determine that the Incore Detector System is OPERABLE.)

SEABROOK - UNIT 1 3/4 3-40 Amendment No. 33

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T GREATER THAN OR E0 VAL TO 350*F m

SVRVEILLANCE RE0VIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position SI-V-3 Accumulator Isolation Open*

SI-V-17 Accumulator Isolation Open*

SI-V-32 Accumulator Isolation Open*

SI-V-47 Accumulator Isolation Open*

SI-V-Il4 SI Pump to Cold-Leg Isolation Open RH-V-14 RHR Pump to Cold-Leg Isolation Open RH-V-26 RHR Pump to Cold-Leg Isolation Open RH-V-32 RHR to Hot-Leg Isolation Closed RH-V-70 RHR to Hot-Leg Isolation Closed SI-V-77 SI to Hot-Leg Isolation Closed SI-V-102 SI to Hot-Leg Isolation Closed b.

At least once per 31 days by:

1)

Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high I

points, and j

Verifying)that each valve (manual, power-operated, orin the flow path that is n 2) automatic otherwise secured in position, is in its correct position.

c.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1)

For all accessible areas of the containment prior to establishing primary CONTAINMENT INTEGRITY, and l

2)

At least once daily of the areas affected within containment by containment entry and during the final entry when primary CONTAINMENT INTEGRITY is established.

l 1

Pressurizer pressure above 1000 psig.

4 l

SEABROOK - UNIT 1 3/4 5-5 Amendment No. 30

l EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,y GREATER THAN OR E00AL TO 350*F SURVEILLANCE REQUIREMENTS 4.5.2 (Continued) d.

At least once per 18 months by:

1)

Verifying automatic interlock action of the RHR system from the Reactor Coolant S stem to ensure that with a simulated or actual Reactor Coolant S stem pressure signal greater than or equal to 365 psig, the interlo ks prevent the valves from being opened.

2)

A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted b structural distress or abnormal corrosion. ) y debris a sump components (trash racks, screens, etc.

show no evidence of e.

At least once per 18 months, during shutdown, by:

1)

Verifying that each automatic valve in the flow path actuates to its correct position on (Safety Injection actuation and Automatic Switchover to Containment Sump) test signals, and 2)

Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a)

Centrifugal charging pump, b)

Safety Injection pump, and c)

RHR pump.

f.

By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1)

Centrifugal charging pump, 2 2480 psid; 2)

Safety Injection pump, 2 1445 psid; and 3)

RHR pump, 2 171 psid.

l l

i I

SEABROOK - UNIT 1 3/4 5-6 Amendment No. 33

}

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,,; GREATER THAN OR E0 VAL TO 350*F SVRVEILLANCE RE0VIREMENTS 4.5.2 (Continued) g.

By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:

1)

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and 2)

At least once per 18 months.

Hiah Head SI System Intermediate Head SI System Valve Number Valve Number SI-V-143 SI-V-80 SI-V-147 SI-V-85 SI-V-151 SI-V-104 SI-V-155 SI-V-109 SI-V-117 SI-V-121 SI-V-125 SI-V-129 h.

By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1)

For centrifugal charging pump lines, with a single pump running:

a)

The sum of the injection line flow rates, excluding the hic hest flow rate, is greater than or equal to 306 gpm, l

anc i

b)

The total pump flow rate is less than or equal to 549 gpm. l 2)

For Safety Injection pump lines, with a single pump running:

a)

The sum of the injection'line flow rates, excluding the highest flow rate, is greater than or equal to 419 gpm, l

and b)

The total pump flow rate is less than or equal to 669 gpm. l 3)

For RHR pump lines, with a single pump running, the sum of the

)

injection line flow rates is greater than or equal to 4213 gpm. l SEABROOK - UNIT 1 3/4 5-7 Amendment No. 33

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T LESS THAN 350*F m

LIMITING CONDITION FOR OPERATION 3.5.3.1 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump, b.

One OPERABLE RHR heat exchanger, c.

One OPERABLE RHR pump, and d.

An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODE 4.

ACTION:

a.

With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal char water storage tank, ging pump or the flow path from the refueling restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T less than 350*F by use of alternate heat removal methods.

m c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

SEABROOK - UNIT 1 3/4 5-8

j 1

3/4 REACTIVITY CONTROL SYSTEMS l

BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that:

(1) tne reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of l

fuel depletion, RCS boron concentration, and RCS T The most restrictive I

condition occurs at E0L, with T atno-loadoperaYfn.g temperature, and is associated with a postulated st,yeam line break accident and resulting uncon-trolled RCS cooldown.

In the analysis of th S accident, a minimum SHUTDOWN MARGIN as specified in the CORE OPERATING LIMTS REPORT (COLR) is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T less than 200* F, the reactivity transients resulting from a postulated sl,eam line break cooldown are minimal. A SHUTDOWN MARGIN as specified in the Col.R and a boron concentration of greater than 2000 ppm are required to permit selficient time for the operater to terminate an inadvertent boron dilution event with T,,less than 200* F.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (HTC) are provided l

l to ensure that the value of this coefficient remains within the limiting l

condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.

These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting end of cycle life (E0L) MTC value as specified in the COLR. The 300 ppm surveillance limit MTC value as specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value as specified in the COLR.

l SEABROOK - UNIT I B 3/4 1-1 Amendment No. 9

,1 REACTIVITY CONTROL SYSTEMS BASES B0 RATION CONTROL 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (Continued)

The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, the MTC is measured as required by Surveillance Requirement 4.1.1.3.a.

A measurement bias is derived from the difference between test measurement and test prediction. All predicted values of MTC for the cycle are conservatively corrected based on measurement bias. The corrected predictions are then compared to the maximum upper limit of Technical Specification 3.1.1.3.

Control rod withdrawal limits are established, if required, to assure all corrected values of predicted MTC will be less positive than the maximum upper limit required by Technical Specification 3.1.1.3.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551' F.

This.

limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3 the OPERABLE status with a steam bu)bble, pressurizer is capable of being in anand (4) the re minimum RT, temperature.

3/4.1.2 B0 RATION SYSTF35 The Boron Ir.jection System ensures that negative reactivity control is available during reach mode of facility operation. The com perform this function include:

(1) borated water sources,ponents required to power supply fron OPERAELE d)iesel generators. boric acid tr (3) separate flow paths (4

(5 an emergency With the RCS in MODES 1, 2 or 3, a minimum of two boron injection flow paths are required to ensure sin le functional capability in the event an assumed failure renders one of t e flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT from ex conditions after xenon decay and cooldown to 200* F. pected operating The maximum expected boron capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 22,000 gallons of 7000 ppm borated water from the boric acid storage tanks or a minimum contained volume of 477,000 gallons of 2000 ppm borated water from the refueling water storage tank (RWST).

The limitation for a maximum of one centrifugal chargin pump to be OPERABLE and the Surveillance Requirement to verify all charg ng pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, an 6 provides assurance that a mass addition pressure transient can be relieved by operation of a single PORY or an RHR suction relief valve.

As a result of this, only one boron injection system is available. This is acceptable on the basis of the stable reactivity condition of the reactor, the emergency power supply requirement for the OPERABLE charging pump and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity l

changes in the event the single injection system becomes inoperable.

I SEABROOK - UNIT 1 B 3/4 1-2 Amendment No. 33

~

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency 1 events b :

(1) maintaining the minimum DNBR in the core greater than or equal to the desi n DN8R value during normal operation and in short-term transients, and l

mechanical properties to within assumed design criteria.(2) limi ing the fission gas r and cladding In a the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux a

on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the integral of linear power along the rod with the highest integrated power i

to the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) specified in the CORE OPERATING LIMITS REPORT (COLR) assure that the design limits on peak local power density and minimum DNBR are not exceeded during normal operation and the consequences of any Non-LOCA event would be within specified acceptance criteria.

For operation with the Fixed Incore Detectors FIDS), assurance that the or i)n the event of xenon redistribution following power chang F (Z a

separate Fixed Incore Detector Alarm through the plant process computer. A FIDS Alarm will be generated when a predetermined number of individual detectors exceed their alarm setpoint. The setpoint for each individual detector is adjusted by the normal 5.21% for system measurement uncertainty and 3% for engineering uncertainty.

This assures that the consequences of a LOCA would be within specified acceptance criteria.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer determines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the limits specified in the COLR.

These alarms are active when power is greater than 50% of RATED THERMAL POWER.

SEABROOK - UNIT 1 8 3/4 2-1 Amendment No. 33

=

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL. FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot minimum DNBR are not exceeded (an)dthe design limits on peak local power density and channel factor ensure that:

1 2 in the event of a LOCA, the peak fuel clad temperaturewillnotexceedthe220}FECCSacceptancecriterialimit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single grou) move together with no individual rod insertion differing by more t1an i 12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and

d. The axial power distribution, expressed in terms of AXIAL FLUX l

DIFFERENCE, is maintained within the limits.

1 SEABROOK - UNIT I B 3/4 2-2 Amendment No. 33

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FlVX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

F" above are, will be maintained within its limits provided Conditions a. through d.

maintained.

The design limit DNBR includes margin to offset any rod bow penalty. Margin is also maintained between the safety analysis limit DNBR and the design limit DNBR. This margin is available for plant design flexibility.

When an F, measurement is taken, an allowance for both measurement error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the movable incore detectors, while 5.21% is appropriate for surveillance results determined with the fixed incore detectors. A 3%

allowance is appropriate for manufacturing tolerance.

For operation with the Fixed Incore Detector System (FIDS) Alarm OPERABLE, the cycle-dependent normalizea axial peaking factor, K(Z), specified in COLR accounts for axial power shape sensitivity in the LOCA analysis. Assurance that the F (Z) limit on Specification 3.2.2 is met during both normal operation and in o

the event of xenon redistribution following power changes is provided by the FIDS Alarm through the plant process computer. This assures that the consequences of a LOCA would be within specified acceptance criteria.

For operation with the FIDS Alarm inoperable, the cycle-dependent normalized axial peaking factor, K(Z), specified in COLR accounts for possible xenon redistribution following power changes in addition to axial power shape sensitivity in the LOCA analysis. This assures that the consequences of a LOCA would be within specified acceptance criteria.

When RCS F" is measured, no additional allowances are necessary prior to comparison with Me established limit. A bounding measurement error of 4.13% for F"a has been allowed for in determination of the design DNBR value.

3/4.2.4 OVADRANT POWER TILT RATIO The purpose of this specification is to detect gross changes in core power distribution between monthly Incore Detector System nrveillances. During normal operation the QUADRANT POWER TILT RATIO is set equal to zero once acceptability of core peaking factors has been established by review of incore surveillances. The limit of 1.02 is established as an indication that the power distribution has changed enough to warrant further investigation.

SEABROOK - UNIT 1 B 3/4 2-3 Amendment No. 33

i POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the updated FSAR assumptions and have been analytically demonstrated adequate to assure compliance with acceptance criteria for each analyzed transient. Operating procedures include allowances for measurement and indication uncertainty so that the limits of 594.3*F for T,, and 2185 psig for pressurizer pressure are not exceeded.

RCS flow must be greater than or equal to,1) the Thermal Design Flow (TDF) with an allowance for measurenent uncertainty and, 2) the minimum measured flow used in place of the TDF in the aralysis of DNB related events when the Revised Thermal Design Procedure (RTDP) methodology is utilized.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the specified limit.

1 i

SEABRDOK - UNIT I B 3/4 2-4 Amendment No. 33 j

INSTRUMENTATION l

BASES 3/4.3.1 ind 3/4.3.2 REACTOR TRfP SYSTEM and ENGINEERED SAFETY FEATVRES ACTUATION SYSTEM INSTRUMENTATION (Continued)

Injection pumps start and automatic valves position, (2) Reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) emergency feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, and f

(11) automatic service water valves position.

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T below Setpoint, prevents the opening of the main feedwatervUveswhichwereclosedbyaSafetyInjectionorHigh Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped.

Reactor not tripped - prevents manual block of Safety Injection.

P-11 On increasing pressurizer pressure, P-ll automatically reinstates Safety Injection actuation on low pressurizer pressure. On decreasing pressure, P-ll allows the manual block of Safety Injection actuation on low pressurizer pressure, and the manual block of SI and steamline isolation on steamline low pressure. On the manual block of steamline low pressure, manual block of steamline low pressure automatically initiates steamline isolation on steam generator pressure negative rate - high.

P-14 On increasing steam generator water level, P-14 automatically trips the turbine and all feedwater isolation valves; inhibits feedwater control valve modulation; and blocks the start of the startup feed-water pump.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION HONITORING FOR PLANT OPERATIONS l

The OPERABILITY of the radiation monitoring instrumentation for plant l

l operations ensures that:

(1) the associated action will be initiated when the l

radiation level monitored by each channel or combination thereof reaches its l

Setpoint, (2) the specified coincidence logic is maintained, and (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance.

The radiation moniters for plant operations sense radiation levels in selected plant systems and locations and determine whether or not l

predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents l

SEABROOK - UNIT 1 B 3/4 3-3

j INSTRUMENTATION BASES HONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS (Continued) and abnormal conditions. Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.

3/4.3.3.2 INCORE DETECTOR SYSTEM The Incore Detector System consists of either a) fixed detector strings and their associated signal processing, or b) movable incore detectors and their associated signal processing. OPERABILITY may be met by either fixed detectors or movable detectors but not by a combination of both.

The OPERABILITY of the Incore Detector System ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core.

Quarter-core flux maps, as defined i,(Z) or F1 a full incore flux map is used.

For the purpose of measuring F n WCAP-8548, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore detectors may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable.

3/4.3.3.3 SEISMIC INSTRUMENTATION I

l The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capa-bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent with tne recommendations of Regulatory Guide 1.12, " Instrumentation for Earth-quakes," April 1974.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive i

materials to the atmosphere.

This capability is required to evaluate the need for initiating ~ protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN SYSTEM l

The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit safe shutdown of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of Appendix A to 10 CFR Part 50.

SEABROOK - UNIT 1 B 3/4 3-4 Amendment No. 33

4 DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 52.0 psig and a temperature of 296*F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with a zirconium alloy.

Each fuel rod shall have a l nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of 3.15 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 5.0 weight percent U-235.

CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80% silver,15%

indium, and 5% cadmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650*F, except for the pressurizer which is 680'F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,265 cubic feet at a nominal T,, of 588.5'F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

j SEABROOK - UNIT 1 5-9 Amendment No. 33

2 DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

equivalent to less than or equal to 0.95 when flooded with

a. A k,,, ted water, which includes margin for uncertainty in calculation unbora methods and mechanical tolerances with a 95% probability at a 95%

confidence level.

b. A nominal 10.35 inch center-to-center distance between fuel assemblies placed in the storage racks.

5.6.1.2 The new fuel storage racks are designed and shall be maintained with:

a. A k,,, equivalent to less than or equal to 0.95 when flooded with unborated water, which includes margin for uncertainty in calculational methods and mechanical tolerances with a 95% probability at a 95%

confidence level,

b. A k,,, equivalent to less than or equal to 0.98 when aqueous foam moderation is assumed, which includes margin for uncertainty in calculational methods and mechanical tolerances with a 95% probability at a 95% confidence level,

c. A nominal 21 inch center-to-center distance between fuel assemblies placed in the storage racks.

DRAINAGE i

5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 14 feet 6 inches.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1236 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

1 1

SEABROOK - UNIT 1 5-10 Amendment No. 6

4 ADMINISTRATIVE CONTROLS ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 A routine Annual Radioactive Effluent Release Report covering the operation of the station during the previous calendar year of operation shall be submitted by May 1 of each year.

The Annual Radioac',1ve Effluent Release Reports shall include a summary of the quantities of ra91oactive liquid and gaseous effluents and solid waste released from the station as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis f61owing the format of Appendix B thereof.

For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories:

class of solid wastes (as defined by 10 CFR Part 61),

type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement).

The Annual Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospharic stability.* This same report shall include an l

assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

This same report shall a'so include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE B0UNDARY (Figure 5.1-3) during the report period.

All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (0DCM).

The Annual Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year

In lieu of submission with the Annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

SEABROOK - UNIT 1 6-17 Amendment No. 22

ADMINISTRATIVE CONTROLS ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.8.1.4 (Continued)

I to show conformance with 40 CFR Part 190, " Environmental Radiation Protection l

Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM, pursuant to Specifications 6.12 and 6.13, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.14.

It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.

The Annual Radioactive Effluent Release Report shall also include the i

following:

an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.9 or 3.3.3.10, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORTS 6.8.1.5 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.8.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each relt 1 cycle, or prior to any remaining portion of a reload cycle, for the following:

1.

Cycle dependent Overpower AT and Overtemperature AT trip setpoint parameters and function modifiers for operation with skewed axial power profiles for Table 2.2-1 of Specification 2.2.1, 2.

SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3.1.1.1, 3.

SHUTDOWN MARGIN limit for MODE 5 for Specification 3.1.1.2, 4.

Moderator Temperature Coefficient BOL and E0L limits, and 300 ppm surveillance limit for Specification 3.1.1.3, SEABROOK - UNIT 1 6-18 Amendment No. 33

4 ADMINISTRATIVE CONTROLS 6.8.1.6.a. (Continued) 5.

Shutdown Rod Insertion limit for Specification 3.1.3.5, 1

6.

Control Rod Bank Insertion limits for Specification 3.1.3.6, 7.

AXIAL FLUX DIFFERENCE limits for Specification 3.2.1, 8.

Heat Flux Hot Channel Factor, F"l and K(Z) for Specification 3.2.2, 9.

Nuclear Enthalpy Rise Hot Channel Factor, and F"[, for Specification 3.2.3.

The CORE OPERATING LIMITS REPORT shall be maintained available in the Control Room.

6.8.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1. WCAP-10266-P-A, Rev. 2 with Addenda (Proprietary) and WCAP-11524-A (Nonproprietary), "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", August, 1986 Methodology for Specification:

1 3.2.2 Heat Flux Hot Channel Factor

2. WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Nonproprietary), "NOTRUMP:

A Nodal Transient Small Break and General Network Code", August,1985 Methodology for Specification.

3.2.2 Heat Flux Hot Channel Factor

3. YAEC-1363-A, "CASM0-3G Validation," April 1988.

l YAEC-1659-A, " SIMULATE-3 Validation and Verification," September 1988.

Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1, 2, 3, and 4 3.1.1.2 -

SHUTDOWN MARGIN for MODE 5 3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 -

Shutdown Rod Intertion Limit t

3.1.3.6 -

Control Rod Inrection Limits 3.2.1 AXIAL FLUX DIFIL?iNCE 3.2.2 Heat Flus M Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor

4. Seabrook Station Updated Final Safety Analysis Report, Section 15.4.6,

" Chemical and Volume Control System Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant System".

l Methodology for Specifications:

3.1.1.1 SHUTDOWN MARGIN for MODES 1, 2, 3, and 4

)

3.1.1.2 -

SHUTDOWN MARGIN for MODE 5 j

SEABROOK - UNIT 1 6-18A Amendment No. 33

.=.

ADMINISTRATIVE CONTROLS i

6.8.1.6.b. (Continued) l

5. YAEC-1241, " Thermal-Hydraulic Analysis of PWR Fuel Elements Using the CHIC-KIN Code", R. E. Helfrich, March 1981 i

Methodology for Specification:

i 3.2.1 AXIAL FLUX DIFFERENCE i

3.2.2 Heat Flux Hot Channel Factor i

3.2.3 Nuclear Enthalpy Rise Hot Channel Factor i

6. YAEC-1849P, " Thermal-Hydraulic Analysis Methodology Using VIPRE-01 For PWR Applications, " October 1992 Methodology for Specification:

2.2.1 Limiting Safety System Settings 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor

7. YAEC-1854P, " Core Thermal Limit Protection Function Setpoint Methodology For Seabrook Station, " October 1992 Methodology for Specification:

i 2.2.1 Limiting Safety System Settings 3.1.3.5 Shutdown Rod Insertion Limit t

3.1.' 3. 6 -

Control Rod Insertion Limits 3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3' Nuclear Enthalpy Rise Hot Channel Factor l

8. YAEC-1856P, " System Transient Analysis Methodology Using RETRAN for PWR Applications," December 1992 i

Methodology for Specification:

h elting Safety System Settings 2.2.1 3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 Shutdown Rod Insertion Limit 3.1.3.6 -

Control Rod Insertion Limits l

3.2.1 AXIAL FLUX DIFFERENCE i

l 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor j

9.

YAEC-1752, " STAR Methodology Application for PWRs, Control Rod Ejection, l

Main Steam Line Break," October 1990 Methodology for Specification:

l 3.1.1.3 Moderator Temperature Coefficient i

3.1.3.5 Shutdown Rod Insertion Limit I~

Control Rod Insertion Limits 3.1.3.6 3.2.1 AXIAL FLUX DIFFERENCE Heat Flux Hot Channel Factor 3.2.2 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor SEABROOK - UNIT 1 6-18B Amendment No. 33

ADMINISTRATIVE CONTROLS 6.8.1.6.b. (Continued)

10. YAEC-1855P, "Seabrook Station Unit 1 Fixed Incore Detector System Analysis," October 1992 Methodology for Specification:

3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor

11. YAEC-1624P, " Maine Yankee RPS Setpoint Methodology Using Statistical Combination of Uncertainties - Volume 1 - Prevention of Fuel Centerline Melt," March 1988 Methodology for Specification:

3.2.1 AXIAL FLUX DIFFERENCE 3.2.2 Heat Flux Hot Channel Factor 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 6.8.1.6.c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

SEABROOK - UNIT 1 6-18C Amendment No. 33

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ML20078L902

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